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Chemistry of the Reaction of Fabricated and High Burnup Spent UO2 Fuel with Saline Brines

Objective

This work will provide a basis for modelling, bridging over the gap between experimental results and performance assessment for long-term storage of the fuel in a repository in salt formations in case of brine intrusion.
The research program aims at characterization and qualification of the chemical durability of unprocessed high burnup UO2 fuel as a barrier against radionuclide release for disposal sites in salt formations.
The research programme aims at characterization and qualification of the chemical durability of unprocessed high burnup uranium dioxide fuel as a barrier against radionuclide release for disposal sites in salt formations. The reaction behaviour of the fuel with saline brines will be studied as a function of time, temperature, redox potential, and surface area in order to give insight into the corrosion mechanisms and sources of radionuclide release. Additionally, the solubility of unirradiated uranium dioxide in salt brines is studied for comparison with the reaction behaviour of the irradiated material in order to identify radiolysis and burnup effects and in particular to identify and quantify solubility effects in the degradation of the fuel matrix. Eventually, the ongoing work will provide a basis for modelling, bridging the gap between experimental results and performance assessment for long term storage of the fuel in a repository in salt formations in case of brine intrusion.

Methods, based on the sorption and desorption of uranium on silica gel and solution analyses by laser induced fluorescence, have been developed to measure uranium concentrations as low as 1 part per billion (ppb) from the uranium dioxide dissolution test in salt brines. Uranium dioxide dissolution has been studied as a function of leaching time in sequential tests using three particle sizes and unirradiated fuel pellets.
The dissolution of 1 mm sized uranium dioxide particles has been studied in sequential tests under reducing conditions and in air, using Q-brine (a magnesium chloride rich brine) as leachant.
Work programme:

The reaction behaviour of the fuel with saline brines is going to be studied as a function of time, temperature, redox potential, and surface area in order to give insight into the corrosion mechanisms and sources of radionuclide release. Additionally, the solubility of non-irradiated UO2 in salt brines is studied for comparison with the reaction behaviour of the irradiated material in order to identify radiolysis and burnup effects and in particular to identify and quantify solubility effects in the degradation of the fuel matrix.
General preparations, analytical techniques, and sample preparations.
Characterization of the durability of spent UO2-fuel in saturated NaCl brines.
Solubility tests with non-irradiated UO2.
Modelling of the reaction behaviour of spent fuel with salt brines.

Funding Scheme

CSC - Cost-sharing contracts

Coordinator

Forschungszentrum Karlsruhe Technik und Umwelt GmbH
Address

76021 Karlsruhe
Germany

Participants (2)

EMPRESA NACIONAL DE RESIDUOS RADIOACTIVOS S.A.
Spain
Address
7,Emilio Vargas 7
28043 Madrid
UNIVERSITAT POLITECNICA DE CATALUNYA
Spain
Address
Calle Jordi Girona 31
000 Barcelona